ML18087A610

Application for Amend to Licenses DPR-70 & DPR-75 Modifying Fuel Enrichment for Reactor Cores & New & Spent Fuel Racks to 4.05 Weight % U-235.Nonproprietary Analysis Encl.Proprietary Analysis Withheld (Ref 10CFR2.790)

ML18087A610
Person / Time
Site:	Salem
Issue date:	12/03/1982
From:	Liden E Public Service Enterprise Group
To:	Varga S Office of Nuclear Reactor Regulation
Shared Package
ML18087A611	List: ML18087A610
References
NUDOCS 8212210244
Download: ML18087A610 (14)

Text

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e PS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Nuclear Department December 3, 1982 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20014 Attention:

Mr. Steven A. Varga, Chief Operating Reactors Branch 1 Divi~ion of Licensing Gentlemen:

REQUEST FOR AMENDMENT TO FACILITY OPERATING LICENSE DPR-70 AND DPR-75 SALEM GENERATING STATION UNITS NOS. 1 AND 2 DOCKET NOS. ~so-272 AND 50-311 In accordance with the Atomic Energy Act of 1954, as amended, and the regulations thereunder, we hereby transmit copies of our request for amendment and our analyses of the changes to Facility Operating License DPR-70 for Salem Generating Station, Unit No. 1 and DPR-75 for Salem Generating Station, Unit No. 2.

This change will modify the fuel enrichment for the reactor cores and the new and spent fuel racks to 4.05 weight percent u...,.23 5.

This change involves a single safety issue and is deemed not to involve a significant safety hazard and is a Class III amendment as defined by 10CFR 170.22.

A check in the amount of $4,000 is enclosed.

This submittal includes three (3) signed originais and forty (40) copies.

Very truly yours,

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0212210244 e212o3 PDR ADOCK 05000272 p

PDR

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E. A. Liden Manager -

Nuclear Licensing and Regulation Aool

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Enclosures The Energy People 95*2169 (4M) 3*82 J

STATE OF NEW JERSEY SS.

COUNTY OF SALEM COUNTY OF SALEM RICHARD A. UDERITZ, being duly sworn according to law deposes and says:

I am a Vice President of Public Service Electric and Gas Company, and as such, I find the matters set forth in our Request for Amendment dated December 3, 1982, are true to the best of my knowledge, information and belief.

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RICHARD A~UDEITi~

Subscribed and sworn to before me this <g7.!J. day of :J>.Ccetz~>SR., 1982

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My Commission expires on

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• • "-...r PROPOSED CHANGES.

DESIGN FEATURES TECHNICAL SPECIFICATION SALEM UNIT NOS. 1 and 2 Description of Changes Ref. LCR 82-24 Modify the fuel enrichment limit for the reactor core and the new and spent fuel storage racks to 4.05 weight percent U-235.

Reason for Change Allow optimization of fuel cycle costs and fuel utilization by employing higher enrichment for use in split feeds and eighteen month cycle operation.

Safety Evaluation The effects of increased fuel enrichment have been evaluated.

The analysis for fuel stored in the new fuel storage rack is included as Attachment I, while Attachment II is the analysis for the Salem

_sp~nt rack.

Evaluation of higher enriched fuel in the reactor is

~"dorie0n a cycle basis to assure that the reload design is consistent r*~-,_ with the -current accia?nt analysis.

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Attachment I New Fuel Storage Rack Analysis

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1.

NEUTRON MULTIPLICATION FACTOR lF\\ :'.} /-' _,.-.

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Criticality of fuel assemblies in the new fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. This is done by fixing the minimum separation between assemblies to take advantage of neutron absorption in water and stainless steel.

The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 per-cent confidence level that the effective multiplication factor (Keff) of the fuel assembly array will be less than 0.98 as reconunended in ANSI NlS.2-1973.

The following are the conditions that are assumed in meeting this design basis for the Salem new fuel storage racks.

2.

NORMAL STORAGE

a. The fuel assembl~ contains the highest enrichment authorized without any control rods or any noncontained burnable poison and is at its most reactive point in life. The enrichment of the 17xl7 Westing-house standard fuel assembly is 4.5 w/o U-235 with no depletion or fission product buildup.

• The assembly is conservatively modeled with the assembly grid volume removed and no U-234 and U-236 in the fuel pellet.

b.

The array is either infinite in lateral extent or is surrounded by a conservatively chosen reflector, whichever is appropriate for the design.

The nominal case calculation is infinite in lateral and axial extent. Calculations show that the finite rack is less reactive than the nominal case infinite rack. Therefore, the nominal case of an infinite array of cells is a conservative assumption.

I-1 2419F:6 I

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c. Mechanical uncertainties and biases due to mechanical tolerances during construction are treated by either using "worst case" condi-tions or by perfonning sensitivity studies to obtain the appropriate values.

The items included in the analysis are:

-- stainless steel thickness

-- cell ID

-- center-to-center spacing

-- asymmetric assembly position The calculation method uncertainty and bias is discussed in Section 4.

d.

Credit is taken for the neutron absorption in full length stainless steel structural material.

3.

POSTULATED ACCIDENTS Most accident conditions will not result in an increase in Keff of the rack.

An example is the dropping of a fuel assembly on top of the rack (the rack structure pertinent for criticality is not deformed and the assembly has more than eight inches separating it from the active fuel in the rest of the rack which precludes interaction).

However, accidents can be postulated (under flooded conditions) which would increase reactivity such as inadvertent drop of an assembly between the outside periphery of the rack and pool wall. Therefore, for accident conditions, the double contigency principle of ANS Nl6.l-1975 is applied. This states that it is unnecessary to assume two unlikely, independent, concurrent events to ensure protection against a critica-1 ity accident. Thus, for accident conditions, the absence of water* in the storage pool can be assumed as a realistic initial condition since assuming its presence would be a second unlikely event.

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The absence of water in the storage pool guarantees subcriticality for enrichments less than 5 w/o[lJ. Thus any postulated accidents other than the introduction of water into the storage area will not preclude the pool from meeting the Keff ~ 0.98 limit.

Because the most limiting accident is the introduction of moderation into the storage pool, this accident will be considered in detennining the maximum Keff for the storage pool.

For this accident, possible sources of moderation, such as those that could arise during fire fight-ing operations, are included in the analysis. This "optimum moderation" accident is not a problem in new fuel storage racks because physically achievable water densities (caused, for instance, by sprinklers, foam generators or fog noziles) are considerably too low (<< 0.01 gm/cm3) to yield Keff values higher than full density water.

The optimum achievable moderation occurs with water at 1.0 gm/cm3*

Preferential water density reduction between cells (i.e., boiling between eel ls) is prevented by the rack design.

4.

METHOD FOR CRITICALITY ANALYSIS The calculation method and cross-section values are verified by compar-ison with critical experiment data for assemblies similar to those for which the racks are designed.

This benchmarking data is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions which include strong neutron absorbers, large water gaps and low moderator densities.

The design method which ensures the criticality safety of fuel assem-blies in the spent fuel storage rack uses the AMPX system of codesE2,3J for cross-section generation and KENO 1vE4J for reactivity* determination.

The 218 energy group cross-section library E2J that is the comnon starting point for all cross-sections used for the benchmarks and the storage rack is generated from ENDF/B-IV data.

The NITAWL program adds to this library the. self-shielded resonance cross-sections that are I-3 2419F:6

-- +*

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e appropriate for each particular geometry.

The Nordheim Integral Treat-ment is used.

The 218 groups are reduced to 19 groups by energy and spatial weighting of cross-sections using the XSDRNPM[3J program which

• is a one-dimensional S transport theory code.

These multi-group cross-section sets arenthen used as input to KENO IV [4J which is a three-dimensional Monte Carlo theory program designed for reactivity calculations *

. A set of 27 critical experiments has been analyzed using the above method to demonstrate its applicability to criticality analysis and to establish the method bias and variability. The experiments range from water moderated, oxide fuel arrays separated by various materials (Boral, steel and water) that simulate LWR fuel shipping and storage conditions [5,6] to dry, harder spectrum uranium metal cylinder arrays with various interspersed materials [7J (Plexiglas, steel and air) that demonstrate the wide range of applicability of the method.

The results and some descriptive facts about each of the 27 benchmark critical experiments are given in Table 1. The average Keff of the benchmarks is 0.9998 which demonstrates that there is no bias associated with the method.

The standard deviation of the Keff values is 0.0057 ~k. The 95/95 one sided tolerance limit factor for 27 values is 2.26. Thus, there is a 95 percent probability with a 95 percent con-fidence level that the uncertainty in reactivity, due to the method, is not greater than O.Ol~k.

The total uncertainty to be added to a criticality c.1lculation is:

TU = [(ks)2

+ (ks)2.

]1/2 method nom1nal where (ks)* th dis 0.013 as discussed above, (ks)

. 1 is the me o nom1na statistical uncertainty associated with the particular KENO calculation being used.

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5.

CRITICALITY ANALYSIS FOR RACK DESIGN For nonnal operation and using the method in the above section, the*

Keff for the rack is determined in the following manner.

Keff = Knominal + 8mech + 8method +

2 2

1/2

[(ks)nominal + (ks)method]

Where:

K

• l nom1na B mech B method ks 1

nom1na ksmethod

=

=

=

=

=

nominal case KENO Keff Keff bias to account for the fact that mechanical tolerances can result in spacings between assemblies less than nominal method bias determined from benchmark critical compari-sons 95/95 uncertainty in the nominal case KENO Keff 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:

. Keff = 0.9189 + 0.0010 + 0.0 + [(.0062) 2 + (.013)2]1/ 2 =.9343 Since Keff is less than 0.98 including uncertainties at a 95/95 proba-bility/confidence level, the acceptance criteria for criticality is met.

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REFERENCES

1. J. T. Thomas, "Nuclear Safety Guide, 11 NUREG/CR-0095 Rev. 2 (June 1978).

2.

W.E. Ford III, et al, 11A 218-Group Neutron Cross-Section Library in the AMPX Master Interface Fonnat for Criticality S~fety Studies, 11 ORNL/CSD/TM-4 (July 1976).

3.

N.M. Green, et al, 11AMPX: A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B, 11 ORNL/TM-3706 (March 1976).

4.

L.M. Petrie and N.F. Cross, "KENO IV-An Improved Monte Carlo Criti-ca 1 i ty Program, 11 ORNL-4938 (November 1978) o

5.

S.R. Biennan, et al, "Critical Separation Between Subcritical Clusters of 2.35 wt % 235uo2 Enriched uo2 Rods in Water with Fixed Neutron Poi sons, 11 Battel le Pacific Northwest Laboratories PNL-2438 (October 1977).

6. S.R. Bierman, et al, "Critical Separation Between Subcritical Cl us~_ers of 4. 29 wt % 235uo2 Enriched uo2 Rods in Water with Fixed Neutron Poisons, 11 Battelle Pacific Northwest Laboratories PNL-2614 (March 1978).

7. J. T. Thomas, "Critical Three-Dimensional Arrays of U (93.2) - Metal Cylinders," Nuclear Science and Engineering, Volume 52, pages 350-359 (1973).

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I FIGURE 1 STRUCTURE BARS INTERMEDIATELY SPACED (HOT INCLUDED IN KENO ~DEL)

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2. O"

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0.25" 0.25"-i T

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REFLECTIVE FUEL ASSEHBLY 17 x 17 W STD.

L _________ _J FIGURE I-1 ANGLE IRONS (FULL LENGTH) 8.432" 9.0" 21.0" l l

TABLE l BENCllMARK CRITICAL EXPERIMENTS [4, 5, 6]

General Enrichment Separating Characterizing Descript ton w/o U235 Reflector Material Separation (cm)

Kerr

1.

uo2 rod lattice 2.35 water water 11.92 1.004 +.004

2.

8.39 0.993 +.004

3.

II II 6.39 1.005 +.004

4.

II ti 4.46 0.994 +.004 e

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5.

II stainless steel 10.44 1.005 +.004

6.

II II 11.47 0.992 +.004

7.

7.76 0.992 +.004

8.

II 7.42 1.004 +.004 1-3

9.

II boral 6.34 1.005.:!:.004

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10.

9.03 0.992 +.004 ti t:r:I

11.

5.05 1.001 +.004 H

12.

4.29 water 10.64 0.999 +.005 I

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13.

II II stainless steel 9.76 0.999 +.005

14.

8.08 0.998 +,006

15.

II boral 6.72 0.998 +.005

16.

U metal cyl lnders 93.2 bare air 15.43 0.998 +.003

17.

II paraffin air 23.84 1.006 +.005

18.

bare atr 19.97 1.005 +.003

19.

II II

• paraffin air 36.47 1.001 +.004

20.

II bare air 13.74 1.005 +.003

21.

II paraffin atr 23.48 1.005 +.004

22.

bare plexlglass

15. 74 1.010 !.003

23.

. paraffin plexlglass 24.43 1.006 +.004

24.

h bare plexlglass

21. 74 0.999 +.003

25.

paraffin plex1glass 27.94 0.994 +.005

26.

bare steel

14. 74 1.000 +.003

27.

bare p lex lg lass. steel 16.67 0.996 +.003 Spent Fuel Storage Rack Analysis

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